• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.11-25
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• 2015

The accident at Japan's Fukushima Daiichi nuclear power plant in March 2011, caused by an earthquake and a subsequent tsunami, resulted in a failure of the power systems that are needed to cool the reactors at the plant. The accident progression in the absence of heat removal systems caused Units 1-3 to undergo fuel melting. Containment pressurization and hydrogen explosions ultimately resulted in the escape of radioactivity from reactor containments into the atmosphere and ocean. Problems in containment venting operation, leakage from primary containment boundary to the reactor building, improper functioning of standby gas treatment system (SGTS), unmitigated hydrogen accumulation in the reactor building were identified as some of the reasons those added-up in the severity of the accident. The Fukushima accident not only initiated worldwide demand for installation of adequate control and mitigation measures to minimize the potential source term to the environment but also advocated assessment of the existing mitigation systems performance behavior under a wide range of postulated accident scenarios. The uncertainty in estimating the released fraction of the radionuclides due to the Fukushima accident also underlined the need for comprehensive understanding of fission product behavior as a function of the thermal hydraulic conditions and the type of gaseous, aqueous, and solid materials available for interaction, e.g., gas components, decontamination paint, aerosols, and water pools. In the light of the Fukushima accident, additional experimental needs identified for hydrogen and fission product issues need to be investigated in an integrated and optimized way. Additionally, as more and more passive safety systems, such as passive autocatalytic recombiners and filtered containment venting systems are being retrofitted in current reactors and also planned for future reactors, identified hydrogen and fission product issues will need to be coupled with the operation of passive safety systems in phenomena oriented and coupled effects experiments. In the present paper, potential hydrogen and fission product issues raised by the Fukushima accident are discussed. The discussion focuses on hydrogen and fission product behavior inside nuclear power plant containments under severe accident conditions. The relevant experimental investigations conducted in the technical scale containment THAI (thermal hydraulics, hydrogen, aerosols, and iodine) test facility (9.2 m high, 3.2 m in diameter, and $60m^3$ volume) are discussed in the light of the Fukushima accident.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2720-2727
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• 2022

The Generation IV Lead-cooled fast reactor (LFR) take the liquid lead or lead-bismuth eutectic alloy (LBE) as the coolant of the primary cooling circuit. Combined with the natural characteristics of lead alloy and the design features of LFR, the system is the simplest and the number of equipment is the least, which reflects the inherent safety characteristics of LFR. The nuclear main coolant pump (MCP) is the only power component and the only rotating component in the primary circuit of the reactor, so the various operating characteristics of the MCP are directly related to the safety of the nuclear reactor. In this paper, various working conditions that may occur in the normal rotation (positive rotating) of the MCP and the corresponding internal flow characteristics are analyzed and studied, including the normal pump condition, the positive-flow braking condition and the negative-flow braking condition. Since the corrosiveness of LBE is proportional to the fluid velocity, the distribution of flow velocity in the pump channel will be the focus of this study. It is found that under the normal pump condition and positive-flow braking conditions, the high velocity region of the impeller domain appears at the inlet and outlet of the blade. At the same radius, the pressure surface is lower than the back surface, and with the increase of flow rate, the flow separation phenomenon is obvious, and the turbulent kinetic energy distribution in impeller and diffuser domain shows obvious near-wall property. Under the negative-flow braking condition, there is obvious flow separation in the impeller channel.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.696-706
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• 2023

Consequences of an anticipated Beyond Design Basis Accident (BDBA) Long-Term Station Blackout (LTSBO) event with complete loss of grid power in the VVER-1200 reactor of Rooppur Nuclear Power Plant (NPP) of Unit-1 are assessed using the RASCAL 4.3 code. This study estimated the released radionuclides, received public radiological dose, and ground surface concentration considering 3 accident scenarios of International Nuclear and Radiological Event Scale (INES) level 7 and two meteorological conditions. Atmospheric transport, dispersion, and deposition processes of released radionuclides are simulated using a straight-line trajectory Gaussian plume model for short distances and a Gaussian puff model for long distances. Total Effective Dose Equivalent (TEDE) to the public within 40 km and radionuclides contribution for three-dose pathways of inhalation, cloudshine, and groundshine owing to airborne releases are evaluated considering with and without passive safety Emergency Core Cooling System (ECCS) in dry (winter) and wet (monsoon) seasons. Source term and their release rates are varied with the functional duration of passive safety ECCS. In three accident scenarios, the TEDE of 10 mSv and above are confined to 8 km and 2 km for the wet and dry seasons, respectively in the downwind direction. The groundshine dose is the most dominating in the wet season while the inhalation dose is in the dry season. Total received doses and surface concentration in the wet season near the plant are higher than those in the dry season due to the deposition effect of rain on the radioactive substances.

• Journal of Energy Engineering
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• v.12 no.2
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• pp.74-83
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• 2003

In the advanced nuclear power plants including APR1400, the SDVS (Safety Depressurization and Vent System) is adopted to increase the plant safety using the concept of feed-and-bleed operation. In the case of the TLOFW (Total Loss of Feedwater), the POSRV (Power Operated Safety Relief Value) located at the top of the pressurizer is expected to open due to the pressurization of the reactor coolant system and discharges steam and/or water mixture into the water pool, where the mixture is condensed. During the condensation of the mixture, thermal-hydraulic loads such as pressure and temperature variations are induced to the pool structure. For the pool structure design, such thermal-hydraulic aspects should be considered. Understanding the phenomena of the submerged steam jet condensation in a water pool is helpful for system designers to design proper pool structure, sparger, and supports etc. This paper reviews and evaluates the steam jet condensation in a water pool on the physical phenomena of the steam condensation including condensation regime map, heat transfer coefficient, steam plume, steam jet condensation load, and steam jet induced flow.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1540-1554
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• 2023

To use nuclear energy sustainably, spent nuclear fuel, classified as high-level radioactive waste and inevitably discharged after electricity generation by nuclear power plants, must be managed safely and isolated from the human environment. In Korea, the land area is limited and the amount of high-level radioactive waste, including spent nuclear fuels to be disposed, is relatively large. Thus, it is particularly necessary to maximize disposal efficiency. In this study, a high-efficiency deep geological repository concept was developed to enhance disposal efficiency. To this end, design strategies and requirements for a high-efficiency deep geological repository system were established, and engineered barrier modules with a disposal canister for pressurized water reactor (PWR)-type and pressurized heavy water reactor type Canada deuterium uranium (CANDU) plants were developed. Thermal and structural stability assessments were conducted for the repository system; it was confirmed that the system was suitable for the established strategies and requirements. In addition, the results of the nuclear safety assessment showed that the radiological safety of the new system met the Korean safety standards for disposal of high-level radioactive waste in terms of radiological dose. To evaluate disposal efficiency in terms of the disposal area, the layout of the developed disposal areas was assessed in terms of thermal limits. The estimated disposal areas were 2.51 km² and 1.82 km² (existing repository system: 4.57 km²) and the excavated host rock volumes were 2.7 Mm³ and 2.0 Mm³ (existing repository system: 4.5 Mm³) for thermal limits of 100 ℃ and 130 ℃, respectively. These results indicated that the area and the excavated volume of the new repository system were reduced by 40-60% compared to the existing repository system. In addition, methods to further improve the efficiency were derived for the disposal area for deep geological disposal of spent nuclear fuel. The results of this study are expected to be useful in establishing a national high-level radioactive waste management policy, and for the design of a commercial deep geological repository system for spent nuclear fuels.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.9 no.1
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• pp.15-19
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• 2013

The auxiliary feedwater is an important to remove the heat from the reactor core when the main feedwater system is unavailable. In most initiating events in Probabilistic Safety Assessment(PSA), the operaton of this system is required to mitigate the accidents. For one of domestic nuclear power plants, a design change of a turbine-driven auxiliary feedwater pump(TD-AFWP), pipe, and valves in the auxiliary system is implemented due to the aging related deterioration by long time operation. This change includes the replacement of the TD-AFWP, the relocation of some valves for improving the system availability, a new cross-tie line, and the installation of manual valves for maintenance. The design modification affects the PSA because the system is critical to mitigate the accidents. In this paper, the safety effect of the change of the auxiliary feedwater system is assessed with regard to the PSA view point. The results demonstrate that this change can supply the auxiliary feedwater from the TD-AFWP in the accident with the motor-driven auxiliary feedwater pump(MD-AFWP) unavailable due to test or maintenance. In addition, the change of MOV's normal position from "close" to "open" can deliver the water to steam generator in the loss of offsite power(LOOP) event. Therefore, it is confirmed that the design change of the auxiliary feedwater system reduces the total core damage frequency(CDF).

• Journal of Korean Society of Industrial and Systems Engineering
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• v.21 no.48
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• pp.291-297
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• 1998

This paper considers the determination of the priority of investment on R&D planning on nuclear power technology using AHP. Due to the complexity of problem, the hierarchy was divided into two parts; main hierarchy and sub-hierarchies. The result shows that the priority of technology options of evolutionary reactor and safety evaluation was higher than that of others.

• Structural Engineering and Mechanics
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• v.10 no.1
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• pp.17-26
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• 2000

This paper presents a case study with a multi-degree-of-freedom (MDOF) system where the Floor Response Spectra (FRS) have been derived from a large ensemble of ground motion accelerograms. The FRS are evaluated by the frequency response function which is calculated numerically. The advantage of this scheme over a repetitive time-history analysis of the entire structure for each accelerogram of the set has been highlighted. The present procedure permits generation of FRS with a specified probability of exceedence.

• Journal of Energy Engineering
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• v.25 no.4
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• pp.85-92
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• 2016

Since the power density of the VHTR(Very High Temperature Reactor) is lower, there is less possibility of core melt. VHTR has no risk of explosion caused by hydrogen generation when the loss of coolant accident occurs, which is another advantage. Along with safety benefit, it can be used as a process heat supplier near demand facilities because coolant temperature is very high enough to be used for industrial purpose. In this paper, we designed the primary system using VHTR and the secondary system providing electricity and process heat. Based on that 350 MW thermal reactor proposed by NGNP(Next Generation Nuclear Part), we developed conceptual model that the IHX(Intermediate Heat Exchanger) loop transports 300 MW thermal energy to the secondary system. In addition, we analyzed thermodynamic behavior and performed the efficiency analysis and optimization study depending on major parameters.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1138-1147
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• 2018

Delayed hydride cracking (DHC) is an important failure mechanism for Zircaloy tubes in the demanding environment of nuclear reactors. The threshold stress intensity factor, $K_{IH}$, and critical hydride length, $l_C$, are important parameters to evaluate DHC. Theoretical models of them are developed for Zircaloy tubes undergoing non-homogenous temperature loading, with new stress distributions ahead of the crack tip and thermal stresses involved. A new stress distribution in the plastic zone ahead of the crack tip is proposed according to the fracture mechanics theory of second-order estimate of plastic zone size. The developed models with fewer fitting parameters are validated with the experimental results for $K_{IH}$ and $l_C$. The research results for radial cracking cases indicate that a better agreement for $K_{IH}$ can be achieved; the negative axial thermal stresses can lessen $K_{IH}$ and enlarge the critical hydride length, so its effect should be considered in the safety evaluation and constraint design for fuel rods; the critical hydride length $l_C$ changes slightly in a certain range of stress intensity factors, which interprets the phenomenon that the DHC velocity varies slowly in the steady crack growth stage. Besides, the sensitivity analysis of model parameters demonstrates that an increase in yield strength of zircaloy will result in a decrease in the critical hydride length $l_C$, and $K_{IH}$ will firstly decrease and then have a trend to increase with the yield strength of Zircaloy; higher fracture strength of hydrided zircaloy will lead to very high values of threshold stress intensity factor and critical hydride length at higher temperatures, which might be the main mechanism of crack arrest for some Zircaloy materials.